Corrosion resistant zirconium alloys containing copper, nickel and iron

ABSTRACT

Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor consist essentially of by weight percent about 0.5 to 2.0 percent tin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron, wherein the copper is at least 0.05 percent, and the balance zirconium. Nuclear fuel elements for use in the core of a nuclear reactor have improved corrosion resistant cladding made from these zirconium alloys or composite claddings have a surface layer of the corrosion resistant zirconium alloys metallurgically bonded to the outside surface of a Zircaloy alloy tube. The claddings may contain an inner barrier layer of moderate purity zirconium metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

This application is a division of application Ser. No. 07/356,474, filed5/25/89, now U.S. Pat. No. 4,986,957.

BACKGROUND OF THE INVENTION

1. Field of the Invention

The present invention relates to zirconium based alloys suitable for usein nuclear reactor service, and in particular for use in claddings offuel elements.

2. Description of Related Art

Nuclear fuel element cladding serves several purposes and two primarypurposes are: first, to prevent contact and chemical reactions betweenthe nuclear fuel and the collant or the moderator if a moderator ispresent; and second, to prevent the radioactive fission products, someof which are gases, from being released from the fuel into the coolantor the moderator. The failure of the cladding, i.e., a loss of theleak-proof seal, can contaminate the coolant or moderator and theassociated systems with radioactive long-lived products to a degreewhich interferes with plant operation.

Zirconium-based alloys have long been used in the cladding of fuelelements in nuclear reactors. A desirable combination is found inzirconium by virtue of its low thermal neutron cross-section and itsgenerally acceptable level of resistance to corrosion in a boiling waterreactor environment. Zircaloy 2, a zirconium alloy consisting of about1.2 to 1.7 percent tin, 0.07 to 0.2 percent iron, 0.05 to 0.15 percentchromium, 0.03 to 0.08 percent nickel, up to 0.15 percent oxygen, andthe balance zirconium, has enjoyed performance in reactor service, butalso possesses some deficiencies that have prompted further research tofind materials providing improved performance. For example, Zircaloy 2cladding on fuel elements in nuclear reactors absorbs hydrogen while thereactor is operating. When the reactor is shut down and the claddingcools the Zircaloy 2 is embrittled by the absorbed hydrogen. Zircaloy 4was one alloy developed as a result of further research to improveZircaloy 2. Zircaloy 4 is similar to Zircaloy 2 but contains less nickel(0.007% max. wt. percent) and slightly more iron. Zircaloy 4 wasdeveloped as an improvement over Zircaloy 2 to reduce absorption ofhydrogen in Zircaloy 2. Zircaloy 2 and Zircaloy 4 are herein referred toas the Zircaloy alloys or Zircaloy.

The Zircaloy alloys are among the best corrosion resistant materialswhen tested in water at reactor operating temperatures, typically about290° C., but in the absence of radiation from the nuclear fissionreaction. The corrosion rate in water at 290° C. is very low and thecorrosion product is a uniform, tightly adherent, black ZrO₂ film. Inactual service, however, the Zircaloy is irradiated and is also exposedto radiolysis products present in reactor water. The corrosionresistance properties of Zircaloy deteriorate under these conditions andthe corrosion rate thereof is accelerated.

Research efforts directed at improving the corrosion properties of thezirconium-based alloys have yielded some advances. Corrosion resistancehas been enhanced in some instances through carefully controlled heattreatments of the alloys either prior to or subsequent to materialfabrication. Added heat treatment cycles, however, generally increasethe expense of making finished products, and in those instances where aninstallation requires welding to be performed, the area affected by theheat of the welding operation may not possess the same corrosionresistance characteristics as the remainder of the article. Variationsin the alloying elements employed and the percentages of the alloyingelements have also been propounded in an effort to address thedeterioration in the corrosion-resistance of these alloys when they areirradiated.

The deterioration under actual reactor conditions of the corrosionresistance properties of Zircaloy is not manifested in merely anincreased uniform rate of corrosion. Rather, in addition to the blackZrO₂ layer formed, a localized, or nodular corrosion phenomenon has beenobserved in some instances on Zircaloy tubing in boiling water reactors.In addition to producing an accelerated rate of corrosion, the corrosionproduct of the nodular corrosion reaction is a highly undesirable whiteZrO₂ bloom which is less adherent and lower in density than the blackZrO₂ layer.

The increased rate of corrosion caused by the nodular corrosion reactionwill be likely to shorten the service life of the tube cladding, andalso this nodular corrosion will have a detrimental effect on theefficient operation of the reactor. The white ZrO₂ being less adherent,may be prone to spalling or flaking away from the tube into the reactorwater. On the other hand, if the nodular corrosion product does notspall away, a decrease in heat transfer efficiency through the tube intothe water is created when the nodular corrosion proliferates and theless dense white ZrO₂ covers all or a large portion of a tube.

Actual reactor conditions cannot be readily duplicated for normallaboratory research due to the impracticality of employing a radiationsource to simulate the irradiation experienced in a reactor.Additionally, gaining data from actual use in reactor service is anextremely time consuming process. For this reason, there is noconclusory evidence in the prior art which explains the exact corrosionmechanism which produces the nodular corrosion. This limits, to somedegree, the capability to ascertain whether other alloys will besusceptible to nodular corrosion before actually placing samples madefrom these alloys into reactors.

Laboratory tests conducted under the conditions normally experienced ina reactor at approximately 300° C. and 1000 psig in water, but absentradiation, will not produce a nodular corrosion product on Zircaloyalloys like that found in some instances on Zircaloy alloys which havebeen used in reactor service. However, if steam is used, with thetemperature increased to over 500° C. and the pressure raised to 1500psig, a nodular corrosion product like that occasionally found onZircaloy in reactor service can be produced on Zircaloy alloys inlaboratory tests. Specimens of Zircaloy alloys which are annealed at750° C. for 48 hours are particularly susceptible to nodular corrosionunder these test conditions. These annealed Zircaloy specimens willproduce, in tests run for relatively short times, i.e. 24 hours, adegree of nodular corrosion comparable to that of Zircaloy tube claddingin actual reactor service that has been found to have nodular corrosion.At this higher temperature and pressure, a simulated nuclear reactorenvironment is provided which will allow researchers to determine thesusceptibility of new alloys to nodular corrosion. With this test, acomparison between samples from new alloys and Zircaloy specimens testedunder the same conditions can be made.

To be considered as a suitable alternate or replacement for the Zircaloyalloys, any new alloy must not only be less susceptible than theZircaloy alloys to nodular corrosion, but must maintain acceptableuniform corrosion rates, comparable to those of the Zircaloy alloys, toensure sufficient service life. Zircaloy alloys have been usedextensively as fuel rod cladding and are known to contain many desirableproperties that alternate or replacement alloys must also contain.Zircaloy alloys have the desirable properties of a low neutronabsorption cross section and at temperatures below 750° F. are strong,ductile, extremely stable and as mentioned previously have excellentuniform corrosion resistance in water at reactor operating temperatures.

Fuel element performance has revealed another problem with brittlesplitting of nuclear fuel element cladding due to the combinedinteractions between the nuclear fuel, the cladding and the fissionproducts produced during nuclear fission reactions. It has beendiscovered that this undesirable performance is due to localizedmechanical stresses on the fuel cladding resulting from differentialexpansion and friction between the fuel and the cladding. Fissionproducts are created in the nuclear fuel by the fission chain reactionduring operation of a nuclear reactor, and these fission products arereleased from the nuclear fuel and are present at the cladding surface.These localized stresses and strains in the presence of specific fissionproducts, such as iodine and cadmium, are capable of producing claddingfailures by phenomena known as stress corrosion cracking or liquid metalembrittlement.

SUMMARY OF THE INVENTION

The present invention relates to corrosion resistant zirconium-basedalloys and corrosion resistant nuclear fuel elements encased withcladding container tubing made from such corrosion resistant zirconiumalloys. In one embodiment, a corrosion resistant first alloy consistsessentially of by weight percent about 0.5 to 2.0 percent tin, about0.24 to 0.40 percent of a solute composed of copper, nickel and ironwherein the copper is at least 0.05 percent, and the balance zirconium.

In another embodiment, a corrosion resistant second alloy consistsessentially of in weight percent about 0.5 to 2.0 percent tin, a solutecomposed of copper, iron and nickel so that each solute element ispresent in an amount from 0.05 to 0.20 percent, and the balancezirconium.

In another embodiment, a corrosion resistant third alloy consistsessentially of in weight percent about 0.5 to 2.0 percent tin, about0.25 to 0.35 percent of a solute composed of copper and nickel whereinthe copper is at least 0.05 percent, and the balance zirconium.

These alloys provide increased resistance to nodular corrosion in highpressure and temperature steam testing, and will maintain acceptableuniform corrosion rates in water and steam tests.

Corrosion-resistant nuclear fuel elements are provided by makingelongated cladding containers from the first, second or third zirconiumalloys described above.

Improved corrosion resistant nuclear fuel elements are also made fromcomposite cladding container tubing having a Zircaloy alloy tube with asurface layer metallurgically bonded on the outside of the Zircaloytube. The surface layer being about 5 to 20 percent of the thickness ofthe Zircaloy tube and consisting essentially of the first, second, orthird zirconium alloy described above. The surface layer is a protectiveshield thick enough to prevent nodular corrosive attack on the Zircaloytube.

Another nuclear fuel element is made from a composite cladding containerthat is resistant to nodular corrosion, stress corrosion cracking andliquid metal embrittlement. An elongated composite cladding container ismade from a Zircaloy alloy tube having a corrosion resistant surfacelayer metallurgically bonded to the outside surface and an inner barrierlayer of zirconium metallurgically bonded on the inside of the alloytube. The inner barrier being about 1 to 30 percent of the thickness ofthe Zircaloy tube and comprised of moderate purity zirconium such assponge zirconium. The outer surface layer being about 5 to 20 percent ofthe thickness of the Zircaloy tube, and consisting essentially of thefirst, second or third zirconium alloy described above.

Cladding container tubing is manufactured by heating an extrusion billetof the first, second or third zirconium alloy described above to about590° to 650° C., extruding the billet into tube shell followed bystandard tube reduction and subsequent heat treatments at about 570° to590° C. to achieve desired tube dimensions and mechanical properties.The standard tube reduction process of zirconium alloy tubing used innuclear fuel elements is pilger-rolling. Pilger-rolling is a tubereduction process using traveling, rotating dies on the outer tubesurface to forge the tube over a stationary mandrel die inside the tube.

Composite cladding containers are manufactured by starting with a tubeblank made from a Zircaloy alloy, and an outer tube placed on this tubeblank. The outer tube is composed of the first, second or thirdzirconium alloys described above. This composite tube is then heated toa temperature in the range of 590° to 650° C. and is extruded. In theprocess, a metallurgical bond between the two zirconium alloys results.Subsequent tube reduction and heat treatments between 570° to 590° C.are performed to achieve the desired tube dimensions and mechanicalproperties. The outer tube is of at least a thickness so that after tubereduction it is about 5 to 20 percent of the thickness of the Zircaloytube.

Another composite cladding container is manufactured by starting with atube blank of a Zircaloy alloy and an outer tube is placed on the tubeblank. The outer tube is made from the first, second or third zirconiumalloys described above. A hollow collar of a metal barrier is placedinside the tube blank. The metal barrier is comprised of moderate purityzirconium such as sponge zirconium. The composite tubing is heated to590° to 650° C. and extruded to form a metallurgical bond between theoutside surface layer and the tube blank, and between the inner metalbarrier and the tube blank. Extrusion is followed by tube reduction andsubsequent heat treatments between 570° C. and 590° C. to develop thedesired tube dimensions and mechanical properties. The outer tube andhollow collar are of at least a thickness so that after tube reductionthe outer layer is about 5 to 20% of the thickness of the Zircaloy tubeand the inner barrier is about 1 to 30% of the Zircaloy tube thickness.

The cladding containers and the composite cladding containers enclose anuclear fuel material, leaving a gap between the fuel and the cladding.In the composite cladding containers having a barrier layer, the barrierlayer shields the alloy tube from the nuclear fuel material held in thecladding as well as shielding the alloy tube from fission products andgases. Because of its purity the liner remains soft during irradiationand minimizes localized strain inside the nuclear fuel element, thusserving to protect the alloy tube from stress corrosion cracking orliquid metal embrittlement.

The first, second or third zirconium alloys and the barrier layer do notintroduce any significant neutron capture penalites, heat transferpenalites, or material incompatibility problems for the nuclear fuelelements of this invention.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 is side view of a partial cross section of a nuclear fuelassembly.

FIG. 2 is a cross section of a nuclear fuel rod cladding containingnuclear fuel.

FIG. 3 is a cross section of a nuclear fuel rod cladding having an innerbarrier layer, and containing nuclear fuel.

FIG. 4 is a cross section of a nuclear fuel rod cladding having an outersurface layer, and containing nuclear fuel.

FIG. 5 is a cross section of a nuclear fuel rod cladding having an innerbarrier layer and an outer surface layer, and containing nuclear fuel.

DETAILED DESCRIPTION OF THE INVENTION

The alloys of the present invention have demonstrated adequateresistance to uniform corrosion to be considered for nuclear reactorservice, and the corrosion resistance is generally comparable to theexcellent uniform corrosion resistance possessed by the Zircaloy alloys.The alloys of the present invention also demonstrate improved resistanceto nodular corrosion.

The addition of tin to zirconium has been practiced in the art prior tothis invention, as evidenced by Zircaloy and other known zirconium-basedalloys. The presence of tin, which stabilizes the alpha-form ofzirconium, primarily contributes to the strength of the alloy, althoughthere is some improvement in uniform corrosion resistance attributableto tin. It has been determined that concentration levels below about0.5% by weight of tin will yield an alloy in which the uniform corrosionrate in water is unacceptably high. Additionally, concentration levelsin excess of about 2.0% by weight of tin will yield an alloy whichdemonstrates an unacceptable level of accelerated corrosion inlaboratory steam tests. The alloys of the present invention have a tincontent within the range of from about 0.5 to about 2.0 percent byweight, preferably from 1.0 to about 1.5 percent by weight, and mostpreferably contain about 1.5% by weight of tin. The alloys of thepresent invention further contain certain additional alloying elements,termed collectively herein as a solute portion of the alloy. The soluteportion of the alloys of the present invention differs from theadditional alloying elements found in the Zircaloy alloys, and isprimarily responsible for the comparative increase in resistance tonodular corrosion. Conventional impurities will also be present in thesealloys.

It should be noted that the alloys of the present invention will alsooptionally contain from about 0.09 to 0.16 weight percent of oxygen.Most commercial grade sponge zirconium which would be used in makingalloys such as the ones in the present invention will contain smallamounts of oxygen, roughly on the order of about 800-1300 parts permillion. In some instances, it will be desirable to increase theconcentration of oxygen in the alloy. Adding oxygen is one way toincrease room temperature yield strength. Thus, the alloys of thepresent invention may be produced with or without the additional oxygen,as this will have little or no effect on the corrosion resistance of thealloys.

There are several parameters which should be considered in choosingalloying elements for the candidate zirconium based alloys to be usedfor fuel cladding in boiling water reactor service. The thermal neutroncross-section of the element should be relatively low to permit productsof the fission reaction to easily pass through the fuel cladding,thereby allowing the boiling water reactor to operate as efficiently aspossible. The cost of the material should be taken into account, andmust not be prohibitively high. The ease or difficulty with which analloy containing the element or elements and zirconium can be producedmust also be considered. It is further desired that the element orelements will enhance the corrosion resistance properties of thezirconium under actual or simulated boiling water reactor conditions.

The thermal neutron cross-section of an element is generally a knownproperty of the element if it has ever come under consideration for usein a nuclear reactor. The costs of the materials can be ascertained fromhistoric price data, with extrapolation if required. The alloyingprocess of the alloys of the present invention is similar toconventional methods for alloying zirconium and thus ease of makingalloy additions is fairly predictable. The alloying is accomplishedpreferably by arc melting a zirconium billet having a suitable amount ofthe alloying metals encased in a hollow portion of the billet. Thismolten metal is then cast as an alloy billet, which will then besubjected to finishing processes to produce final shapes.

Generally, of the parameters discussed above, the most difficult topredict is whether the alloying element will contribute to theenhancement of corrosion resistance.

In the present invention, zirconium-based alloys have been discoveredthat perform substantially better than Zircaloy 2 in tests conducted todetermine resistance to nodular corrosion. These alloys also performwell in tests for determining resistance to uniform corrosion. A firstalloy consists essentially of by weight percent 0.5 to 2.0 percent tin,about 0.24 to 0.40 percent of a solute composed of copper, nickel andiron, wherein the copper is at least 0.05 percent, and the balancezirconium. A second alloy consists essentially of by weight percentabout 0.5 to 2.0 percent tin, a solute composed of copper, iron andnickel so that each solute element is present in an amount from 0.05 to0.20 percent, and the balance zirconium. A third alloy consistsessentially of by weight percent about 0.5 to 2.0 percent tin, about0.25 to 0.35 weight percent of a solute composed of copper and nickelwherein the copper is at least 0.05 percent, and the balance zirconium.

Solute elements copper, nickel and iron possess the low thermal neutroncross-section, low cost, ease of alloying, and corrosion resistanceproperties that are desirable in zirconium based alloys.

Tests for both uniform corrosion resistance and nodular corrosion havebeen conducted on alloys of the present invention. These tests haveshown that a dramatic decrease in susceptibility to nodular corrosioncan be attained in an alloy which is relatively insensitive to heattreatment while retaining essentially the same uniform corrosionresistance of a Zircaloy 2 alloy. Solute concentrations ranging from aslow as 0.24 weight percent to as high as 0.40 weight percent have beentested and have been shown to exhibit superior resistance to nodularcorrosion, compared to the performance of Zircaloy 2.

Alloys having copper and nickel as solutes had greatly improvedresistance to nodular corrosion when they were given the 750° C./48 houranneal that sensitizes Zircaloy 2 to nodular corrosion. Zirconium alloytubing is heat treated several times during tube production; therefore,the zirconium alloys containing solutes copper and nickel will provideimproved nodular corrosion resistance when properly heat treated duringtube production.

The corrosion-resistant fuel elements of this invention are shown byreferring now more particularly to FIG. 1, where there is shown apartially cutaway sectional view of a nuclear fuel assembly 10. Thisfuel assembly consists of a tubular flow channel 11 of generally squarecross section provided at its upper end with lifting bale 12 and at itslower end with a nose piece (not shown due to the lower portion ofassembly 10 being omitted) The upper end of channel 11 is open at 13 andthe lower end of the nose piece is provided with coolant flow openings.An array of fuel elements or rods 14 is enclosed in a channel 11 andsupported therein by means of upper end plate 15 and a lower end plate(not shown due to the lower portion being omitted). The liquid coolantordinarily enters through the openings in the lower end of the nosepiece, passes upwardly around fuel elements 14, and discharges at upperoutlet 13 in a partially vaporized condition for boiling reactors or inan unvaporized condition for pressurized reactors at an elevatedtemperature.

The nuclear fuel elements or rods 14 are sealed at their ends by meansof end plugs 18 welded to the cladding 17, which may include studs 19 tofacilitate the mounting of the fuel rod in the assembly. A void space orplenum 20 is provided at one end of the element to permit longitudinalexpansion of the fuel material and accumulation of gases released fromthe fuel material. A nuclear fuel material retainer means 24 in the formof a helical member is positioned within space 20 to provide restraintagainst the axial movement of the pellet column, especially duringhandling and transportation of the fuel element.

The fuel element is designed to provide an excellent thermal contactbetween the cladding and the fuel material, a minimum of parasiticneutron absorption, and resistance to bowing and vibration which isoccasionally caused by flow of the coolant at high velocity.

A nuclear fuel element or rod 14 is shown in a partial section in FIG. 1constructed according to the teachings of this invention. The fuelelement includes a core or central cylindrical portion of nuclear fuelmaterial 16, here shown as a plurality of fuel pellets of fissionable orfertile material positioned within a structural cladding or container17. In some cases, the fuel pellets may be of various shapes, such ascylindrical pellets or spheres, and in other cases, different fuel formssuch as particulate fuel may be used. The physical form of the fuel isimmaterial to this invention. Various nuclear fuel materials may beused, including uranium compounds, plutonium compounds, thoriumcompounds and mixtures thereof. A preferred fuel is uranium dioxide or amixture comprising uranium dioxide and plutonium dioxide.

Referring now to FIG. 2, the nuclear fuel material 16 forming thecentral core of the fuel element 14 is surrounded by a cladding 17. Thecladding container encloses the core so as to leave a gap between thecore and the cladding container during use in a nuclear reactor. Thecladding is comprised of a corrosion-resistant zirconium alloy tube 21.The alloy tube 21 is made from either the first, second or thirdzirconium alloys described above.

It should be noted that the first, second or third zirconium alloysdescribed herein will also optionally contain from about 0.09 to 0.16weight percent of oxygen. Most commercial grade sponge zirconium whichwould be used in making alloys such as the ones in the present inventionwill contain small amounts of oxygen, roughly on the order of about800-1300 parts per million In some instances, it will be desirable toincrease the concentration of oxygen in the alloy. Adding oxygen is oneway to increase room temperature yield strength. Thus, the alloys of thepresent invention may be produced with or without the additional oxygen,as this will have little or no effect on the corrosion resistance of thealloys.

Another embodiment of this invention is shown by referring to FIG. 3.The nuclear fuel material 16 forming the central core of the fuelelement 14 is surrounded by a composite cladding 17. The compositecladding container encloses the core so as to leave a gap 23 between thecore and the cladding container during use in a nuclear reactor. Thecomposite cladding is comprised of a zirconium alloy tube 21 made fromeither the first, second or third zirconium alloy described above. Thealloy tube has bonded on the inside surface thereof a metal barrier 22so that the metal barrier forms a shield between the alloy tube 21 andthe nuclear fuel material held in the cladding. The metal barrier formsabout 1 to about 30 percent of the thickness of the cladding, and iscomprised of a low neutron absorption material, namely, moderate purityzirconium. One moderate purity zirconium is sponge zirconium. The metalbarrier 22 protects the alloy tube portion of the cladding from contactand reaction with gasses and fission products from the nuclear fuel, andprevents the occurrence of localized stress and strain.

The content of the metal barrier of moderate purity zirconium isimportant and serves to impart special properties to the metal barrier.Generally, there is at least about 1,000 parts per million (ppm) byweight and less than about 5,000 ppm impurities in the material of themetal barrier and preferably less than about 4,200 ppm. Of these, oxygenis kept within the range of about 200 to about 1,200 ppm. All otherimpurities are within the normal range for commercial, reactor-gradesponge zirconium.

In another embodiment of this invention, a corrosion-resistant nuclearfuel element is shown by referring to FIG. 4. The nuclear fuel material16 forming the central core of fuel element 14 is surrounded by acomposite cladding 17. The composite cladding container encloses thecore so as to leave a gap 23 between the core and the cladding containerduring use in a nuclear reactor. The composite cladding is comprised ofa zirconium alloy tube 30 made from a Zircaloy alloy. The alloy tube hasbonded on the outside surface thereof a metal layer 32 so that the metallayer forms a corrosion protective shield over the alloy tube. The outermetal layer is about 5 to 20 percent of the thickness of the alloy tubeand is comprised of either the first, second or third zirconium alloydescribed above. The outer metal layer protects the Zircaloy alloy tubeportion of the cladding from nodular corrosion.

Another improved nuclear fuel element is shown by referring to FIG. 5.The nuclear fuel material 16 forming the central core of fuel element 14is surrounded by a composite cladding 17. The composite claddingcontainer encloses the core so as to leave a gap 23 between the core andthe cladding container during use in a nuclear reactor. The compositecladding is comprised of a Zircaloy alloy tube 30. The alloy tube hasbonded on the inside surface thereof a metal barrier 22 so that themetal barrier forms a shield between the alloy tube 30 and the nuclearfuel material held in the cladding. The metal barrier is about 1 toabout 30 percent of the thickness of the alloy tube and is comprised ofa low neutron absorption material, namely, moderate purity zirconium asdescribed above. The metal barrier 22 protects the alloy tube portion ofthe cladding from contact and reaction with gases and fission productsfrom the nuclear fuel, and prevents the occurrence of localized stressand strain. An outer surface layer is bonded on the outside surface ofthe alloy tube 30. The outer metal layer is about 5 to 20 percent of thethickness of the alloy tube and is comprised of either the first, secondor third zirconium alloy described above. The outer metal layer protectsthe Zircaloy alloy tube portion of the cladding from nodular corrosion.

Sponge zirconium metal forming the metal barrier in the compositecladding is highly resistant to radiation hardening, and this enablesthe metal barrier after prolonged irradiation to maintain desirablestructural properties such as yield strength and hardness at levelsconsiderably lower than those of conventional zirconium alloys. Ineffect, the metal barrier does not harden as much as conventionalzirconium alloys when subjected to irradiation, and this together withits initially low yield strength enables the metal barrier to deformplastically and relieve pellet-induced stresses in the fuel elementduring power transients. Pellet induced stresses in the fuel element canbe brought about, for example, by swelling of the pellets of nuclearfuel at reactor operating temperatures (300° C. to 350° C.) so that thepellet comes into contact with the cladding.

It has further been discovered that a metal barrier of sponge zirconiumof the order preferably about 5 to 15 percent of the thickness of thecladding and a particularly preferred thickness of 10 percent of thecladding bonded to the alloy tube of a zirconium alloy provides stressreduction and a barrier effect sufficient to prevent failures in thecomposite cladding.

The corrosion resistant nuclear fuel rod cladding used in the nuclearfuel elements of this invention can be fabricated from a billetcomprised of a zirconium alloy made from either the first, second orthird zirconium alloy described above. The billet is heated to 590° to650° C. and extruded. The extruded tubing is then subjected to a processinvolving conventional tube reduction until the desired size of tubingis achieved.

In another method, a hollow collar of the sponge zirconium selected tobe the metal barrier is inserted into a hollow billet of the first,second or third zirconium alloy described above. The assembly is heatedto 590° to 650° C. and extruded. The extruded tubing is then subjectedto a process involving conventional tube reduction until the desiredsize of cladding is achieved.

In another method, a tube blank is made from a Zircaloy alloy and anouter tube is placed on this tube blank. The outer tube is composed ofthe first, second or third zirconium alloy described above. Thisassembly is then heated to a temperature in the range of 590° to 650° C.and is extruded. The extruded tubing is then subjected to a processinvolving conventional tube reduction until the desired size of tubingis achieved.

In another method, a tube blank is made from a Zircaloy alloy and anouter tube is placed on this tube blank. The outer tube is composed ofthe first, second or third zirconium alloy described above. A hollowcollar of the sponge zirconium selected to be the metal barrier isinserted into the tube blank. The assembly is heated to a temperature inthe range of 590° to 650° C. and is extruded. The extruded tubing isthen subjected to a process involving conventional tube reduction untilthe desired size of tubing is achieved.

Intermediate and final anneals are used during the tube reductionprocesses described above. Anneals range between 570° to 590° C.

The invention includes a method of producing a nuclear fuel elementcomprising making a cladding or a composite cladding container comprisedof a zirconium alloy, or a zirconium alloy and a barrier layer, or aZircaloy alloy and a surface layer, or a Zircaloy alloy and an outersurface layer and an inner barrier layer. The container is open at oneend and filled with a core of nuclear fuel material leaving a gapbetween the core and the container and leaving a cavity at the open end.A nuclear fuel material retaining means is inserted into the cavity andan enclosure is applied to the open end of the container, leaving thecavity in communication with the nuclear fuel. The end of the cladcontainer is then bonded to the enclosure to form a tight sealtherebetween.

The present invention offers several advantages promoting a longoperating life for a nuclear fuel element. A greater resistance tonodular corrosion protects the strength and integrity of the cladding.On cladding having a barrier layer, the reduction of chemicalinteraction on the cladding, the minimization of localized stress on thezirconium alloy tube portion of the cladding, and the minimization ofstress corrosion and strain corrosion on the zirconium alloy tubeportion of the cladding, all reduce the probability of a splittingfailure occurring in the zirconium alloy tube. The invention furtherreduces expansion or swelling of the nuclear fuel into direct contactwith the zirconium alloy tube, and this reduces the occurrence oflocalized stress on the zirconium alloy tube, initiation or accelerationof stress corrosion of the alloy tube and bonding of the nuclear fuel tothe alloy tube.

An important property of the composite cladding of this invention isthat the foregoing improvements are achieved with no substantialadditional neutron absorption penalty. Further, the composite claddinghas a very small heat transfer penalty in that there is no thermalbarrier to transfer of heat, such as results in a situation where aseparate foil or a liner is inserted in a fuel element. Also, thecomposite cladding of this invention is inspectable by conventionalnon-destructive testing methods during various stages of fabrication andoperation.

The following examples are offered to further illustrate the improvednodular corrosion resistance of the alloys used in this invention.

EXAMPLE I

Table I lists several examples of alloys of the present invention, alongwith three entries at the bottom of the table which are Zircaloy 2alloys in three different cold-rolled and heat treated states. Thesealloys were tested in water containing 8 ppm oxygen at 288° C. and 1500psig, conditions similar to a reactor operating temperature and pressurebut absent the radiation, to evaluate the resistance to uniformcorrosion.

It can be seen from the results in Table I that the tested alloys of thepresent invention exhibit excellent resistance to uniform corrosion.Although tested for much longer periods of time, the tested alloys ofthis invention exhibited corrosion rates comparable to those of theZircaloy 2 specimens. None of the specimens tested under theseconditions exhibited any sign of the formation of modular corrosionproducts.

                  TABLE I                                                         ______________________________________                                        UNIFORM CORROSION TEST                                                        IN WATER AT 288° C., 1500 psig, 8 ppm OXYGEN                                                       Weight Gain                                                                            Elapsed Time                             Sn    Cu      Ni     Fe     (mg/dm.sup.2)                                                                          (Hours)                                  ______________________________________                                        1.5   0.1     0.1    0.1    25       14,849                                   1.0   0.1     0.1    0.1    25       14,849                                   2.0   0.1     0.1    0.1    25       14,849                                   1.59  0.11    0.11   0.13   27       13,309                                   1.61  0.13    0.13   0.14   21       5,077                                    Z2 (Zircaloy 2, cross-rolled                                                                    11.0       1,000                                            commercial plate) 13.2       1,700                                            Z2 with 750° C./16 hr. anneal                                                            11.0       1,000                                                              15.0       1,700                                            Z2 with Beta Quench                                                                             15.0       1,000                                                              17.0       1,700                                            ______________________________________                                    

EXAMPLE II

Table II reports the results of tests conducted to determine thesusceptibility of the alloys of the present invention to nodularcorrosion. The tests were performed using steam at 510° C. and 1500psig. In the laboratory, these same test conditions induce the formationof the nodular corrosion product on Zircaloy alloys which have beengiven a 750° C./48 hour anneal, and is also identical to the nodularcorrosion found on sometimes Zircaloy after being used in reactorservice. For purposes of comparison, the weight gains of annealedZircaloy samples under the same test condition were on the order ofseveral thousand milligrams per square decimeter.

The tests in Table II were also performed on samples in variouscold-reduced and heat treated states. The results in Table II provide anindication that the corrosion resistant properties of these alloys arerelatively insensitive to the heat treatment state of the specimen.However, alloys having the solute composed of copper and nickelpreferably are heat treated. Some compositions were tested usingspecimens in cold-rolled plate form, both with and without a subsequentanneal. Two compositions were tested only after having been cold-reducedand annealed. The 750° C. anneal for 48 hours, which all of the testedsamples were subjected to, is the heat treatment which strips theZircaloy 2 alloy of its resistance to nodular corrosion under thelaboratory steam tests.

All of the weight gains recorded in Table II are far superior to theresults obtained when sensitized Zircaloy 2 annealed at 750° C. for 48hours is tested. Most of the tested alloys of the present inventionproduced weight gains of less than 100 milligrams per square decimeter,and one produced a weight gain of 107 mg/dm². As previously mentioned,weight gains reported in tests of sensitized Zircaloy specimens underthe same test times and conditions are in the order of several thousandmilligrams per square decimeter.

In addition to the reduced weight gains evidenced in the alloys of thepresent invention, none of these alloys showed any sign of formation ofnodular corrosion products. Under the test conditions, these alloysclearly provide improvement in resistance to nodular corrosion.

                  TABLE II                                                        ______________________________________                                        WEIGHT GAIN AFTER EXPOSURE TO                                                 STEAM AT 510° C., 1500 psig FOR 24 HOURS                                                 Weight Gain (mg/dm.sup.2)                                   Zr Alloy Composition                                                                            Cold Rolled 0.1" Plate                                      Weight/Percent               750° C./48 hr.                            Sn    Cu      Ni      Fe    As Rolled                                                                              Anneal                                   ______________________________________                                        1.5   0.2     0.1           865      87                                       1.5   0.1     0.2           367      107                                      1.5   0.1     0.1     0.1   58       70                                       1.5   0.08    0.08    0.08  61       76                                       1.5   0.12    0.12    0.12  53       69                                       1.0   0.1     0.1     0.1   52       69                                       2.0   0.1     0.1     0.1   60       81                                       1.61  0.13    0.13    0.14  *        78                                       1.59  0.11    0.11    0.13  *        65                                       ______________________________________                                         * -- Not tested.                                                         

What is claimed is:
 1. A corrosion resistant alloy, consistingessentially of by weight percent about 0.5 to 2.0 percent tin, about0.24 to 0.40 percent of a solute comprising copper, nickel and iron,wherein the copper is at least 0.05 percent, and the balance zirconium.2. The corrosion resistant alloy according to claim 1, furtherconsisting essentially of 0.09 to 0.16 weight percent oxygen.
 3. Acorrosion resistant alloy consisting essentially of by weight percentabout 0.5 to 2.0 percent tin, a solute comprising copper, iron andnickel so that each solute element is present in an amount of from 0.05to 0.20 percent, and the balance zirconium.
 4. The corrosion resistantalloy of claim 3, further consisting essentially of 0.09 to 0.16 weightpercent oxygen.
 5. A corrosion resistant alloy consisting essentially ofby weight percent about 0.5 to 2.0 percent tin, about 0.25 to 0.35percent of a solute comprising copper and nickel, wherein the copper isat least 0.05 percent and the balance zirconium.
 6. The corrosionresistant alloy of claim 5, further consisting essentially of 0.09 to0.16 weight percent oxygen.